Search NASA⌕ Search

SEARCH · Search NASA

Results for “RADIOISOTOPE NEUTRON SOURCES”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Neutron Detector Signal Processing to Calculate the Effective Neutron Multiplication Factor of Subcritical Assemblies

This report describes different methodologies to calculate the effective neutron multiplication factor of subcritical assemblies by processing the neutron detector signals using MATLAB scripts. The subcritical assembly can be driven either by a spontaneous fission neutron source (e.g. californium) or by a neutron source generated from the interactions of accelerated particles with target materials. In the latter case, when the particle accelerator operates in a pulsed mode, the signals are typically stored into two files. One file contains the time when neutron reactions occur and the other contains the times when the neutron pulses start. In both files, the time is given by an integer representing the number of time bins since the start of the counting. These signal files are used to construct the neutron count distribution from a single neutron pulse. The built-in functions of MATLAB are used to calculate the effective neutron multiplication factor through the application of the prompt decay fitting or the area method to the neutron count distribution. If the subcritical assembly is driven by a spontaneous fission neutron source, then the effective multiplication factor can be evaluated either using the prompt neutron decay constant obtained from Rossi or Feynman distributions or the Modified Source Multiplication (MSM) method.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Neutron Spectra, Fluence and Dose Rates from Bare and Moderated Cf-252 Sources

A new, stronger 252 Cf source (serial number SR-CF-3050-OR) was obtained from Oak Ridge National Laboratory (ORNL) in 2014 to supplement the existing 252 Cf sources which had significantly decayed. A new instrument positioning track system was designed and installed by Hopewell Designs, Inc. in 2011. The neutron field from the new, stronger 252 Cf source in the modified calibration environment needed to be characterized as well as the modified neutron fields produced by the new source and seven different neutron moderators. Comprehensive information about our 252 Cf source, its origin, production, and isotopic content and decay characteristics needed to be compiled as well. This technical report is intended to address these issues.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Isotope Program Report June FY2016

Isotope Program Monthly Highlights are briefly described. These include data on isotopes shipped, updates on equipment fabrication and testing, a potential new approach for nondestructive measurement of the amount of Cf-252 deposited on a surface, and efforts to recover and purify uranium-234 obtained from old PuBe sources.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Radioisotope production at the Spallation Neutron Source: Design concept of experimental target station

Completion of the Proton Power Upgrade Project for the Spallation Neutron Source (SNS) accelerator at Oak Ridge National Laboratory opens an opportunity to utilize reserve beam power of more than 100 kW for applications beyond neutron production. One of these applications is the production of critical radionuclides. To demonstrate the feasibility of using the reserve beam power to produce radioisotope at SNS, a design concept of a small-scale experimental target station in the Linac Dump area has been developed. This experimental facility will provide isotope yield benchmarking data using protons in the GeV range. It will also enable additional research and development in isotope handling and radiochemical separation. The target station consists of a target module enclosed in a vessel and concrete shielding. Particle transport calculations and thermo-mechanical simulations are used to determine beam parameters, decay time, isotope yield, shielding dimensions, and target design parameters. Calculations verified that the irradiated capsule can be handled manually using hands-off tools and transported to a hot cell in a shielded container for post-irradiation characterizations.

Lee, Yong Joong [ORNL] (ORCID:0000000298381723)↗
DOE OSTI2025

Radioisotope production at the Spallation Neutron Source: Design concept of isotope production target

Upon completion of the Second Target Station (STS) Project in the mid 2030s, the Spallation Neutron Source accelerator at Oak Ridge National Laboratory will deliver a 2.7 MW proton beam to the neutron production targets. In the post-STS phase, the accelerator will have a reserve beam power capacity of at least 100 kW beyond what the two neutron production targets will receive, which could potentially be ramped up to 300 kW. In this paper, a design concept for a radioisotope production target that could utilize 250 kW of the reserve beam power capacity is presented. The target consists of thorium discs encapsulated in 316L austenitic steel shells that are cooled by water. The estimated post-irradiation activity of Ac-225 and Ra-225, critical medical radioisotopes used in targeted alpha therapy cancer treatment, is calculated at the end of bombardment after a 14 day long irradiation time. Thermal and structural analyses are performed on the basis of calculated nuclear heating data. The technical feasibility of a high-power target under a 250-kW beam load with an extremely low duty factor of $3.5\cdot 10^{-6}$ is presented from thermal, structural and fatigue lifetime perspectives.

Lee, Yong Joong [ORNL] (ORCID:0000000298381723)↗
DOE OSTI2016

Estimating the AmLi Neutron Spectrum from Measured Ring Ratios

These are a set of slides on estimating the AmLi neutron spectrum from measured ring ratios. The IAEA uses an AmLi source in the Uranium Neutron Coincidence Collar (UNCL) to verify compliance with nonproliferation treaties. The UNCL requires calibration with known uranium samples. The AmLi spectrum isn’t known well enough to allow simulated calibrations. Alphas lose energy traveling through AmO 2 particle of unknown size. Energy reduction below Li threshold enhances O contribution. Unknown Li matrix material affects neutron production and thermalization. There is large variation in spectra from each element. Other topics covered include: applications, physics considerations, current spectra, measurement overview, measurement results - variation between sources, simulations, spectra fitting, other simulations, and conclusions.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2018

Radioisotope power source dose estimation tool (RPS-DET)

The purpose of this work is to develop a radioisotope power source dose estimation tool (RPS-DET) capable of rapidly predicting accurate neutron and gamma-ray flux/dose implications from various radioisotope power sources (RPSs). RPS-DET can use this information to calculate the flux/dose incident on or absorbed by nearby humans, materials, and electronics over a complex mission lifetime. This simulation tool uses a Matlab user interface to combine common geometry scenarios for traditional and new RPS design configurations to write a particle-transport simulation using Monte Carlo N-Particle (MCNP) 6, which provides results that are automatically post-processed for interactive analysis. The user defines the average age of the PuO2fuel, which determines the appropriate time-dependent gamma-ray and neutron spectra, along with the desired RPS and local environment for the simulation. Results for a given simulation are gamma-ray, neutron, and combined flux spectra, whole-body effective dose for humans, and absorbed dose in silicon for electronics. This work has been benchmarked against measurements of existing RPS units and has been determined to predict neutron and gamma flux, absorbed dose, and dose equivalent within 50% of simulation uncertainties.

Smith, Michael↗
DOE OSTI2022

Fast-neutron/gamma-ray radiography using a broad-energy neutron source

Transmission radiography is a well-established nondestructive examination technique with widespread application to fields ranging from medicine to security. Traditionally, inspection is performed using a single particle type (such as x-rays). However, the information available using a single probe for traditional radiography is limited. The present work evaluates material discrimination via transmission radiography using the attenuation ratio of monoenergetic gamma rays to a broad spectrum of fast neutrons. Here, the method was assessed using an 241 Am/Be radioisotope source that provides both 4.4 MeV gamma rays and a spectrum of fast neutrons up to 12 MeV. A total of 14 object configurations were measured: seven different materials each with two thicknesses (2.5 cm and 5 cm). The ability to distinguish materials was evaluated and shows more significant variation among atomic numbers than for high-energy x-rays alone, making it easier to distinguish between classes such as low-, mid-, and high-Z materials. These results suggest that superior material discrimination is also possible using a combination of monoenergetic gamma rays and broad-spectrum fast neutrons from a variety of nuclear reactions, such as 11 B(d,nγ) 12 C, that could be implemented in future inspection systems.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Characterization of a mixed high-energy spallation neutron–proton field using monoisotopic activation detectors

In this paper, we used monoisotopic Co, Au, and Bi threshold activation detectors to characterize the high-energy ( > 5 MeV) portion of a mixed spallation neutron–proton (n/p) field. The field was produced by irradiating the natural uranium target–blanket subcritical assembly Quinta with 660-MeV protons from Phasotron accelerator at the Joint Institute for Nuclear Research (JINR) in Dubna. Experimental yields of the (n,x) and (p,x) reactions were compared to the calculated reaction rates, which represent the convolution of particle flux and reaction cross section. The flux was simulated using the radiation transport code MCNPX 2.7.0, and the cross sections were extracted from the following three sources: the intranuclear cascade model INCL4.2, the phenomenological spherical optical model from TALYS-1.8, and the evaluated nuclear data file TENDL-2015. We show that, in most cases, the evaluated data provide a closer agreement with the experimental data in comparison with INCL4.2 model, which is an integral part of MCNPX. Finally, our results bring a contribution to those fields of physics research that utilize computer prediction of neutron generation in the energy range of tens to hundreds MeV, such as the medical radioisotope production, radiation shielding of spallation neutron sources, and nuclear transmutation in accelerator-driven systems.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Approach for the isotopic characterization of irradiated start-up sources for disposal at WIPP - 15674

The National Nuclear Security Administration, Global Threat Reduction Initiative (GTRI)/ Off-site Source Recovery Program (OSRP) requires an approved final disposition pathway for sealed irradiated startup sources with uncertain irradiation histories. To create this path a reliable method is required to determine yields of transmutation, fission and activation products contained within the source at the time of disposal. Because of the uncertain histories of these sources, a number of likely histories are modeled with the code ORIGEN in order to establish the bounds of the radioisotope yields. The ORIGEN simulations are confirmed with gamma spectroscopy, specifically examining the activities of several fission and transmutation isotopes. These calculations, coupled with other previously established characterization techniques for transuranic neutron sources, will be used to establish a waste profile that is in compliance with Waste Isolation Pilot Plant (WIPP) waste acceptance criteria. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Neutronic Comparison of New HSA Cobalt Capsule Design to Legacy HSA Cobalt Capsule Design

A comparison of the neutronic properties between the legacy high specific activity (HSA) cobalt capsule design and the new HSA cobalt capsule design was conducted utilizing the full advanced test reactor core MCNP model. The neutron flux and effective neutron absorption cross section for the desired (n,gamma) reaction as well as the undesired reactions (n,2n), (n, alpha), and (n,p) for the cobalt target material for each design was calculated and compared. It was found that the effective neutron absorption cross section for the desired reaction (n,gamma) on the target material that directly produces Co-60 is slightly higher in the new design than the legacy design, and there is little statistical difference between the two target designs for the undesired reactions (n,2n), (n, ), and (n,p). Based on these results, neutronically, there does not appear to be a significant difference between the two capsule designs.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Shielding and Activation Analyses for Beam Test Facility at SNS

The beam test facility (BTF) will be a copy of the accelerator front end of the Spallation Neutron Source (SNS) facility and will be used for test and development of new front-end equipment, for conducting research in accelerator physics (in a further build-out) for the SNS second target station moderator development (tests). The BTF is assembled and will be operating in the radio frequency (RF) on-site annex building. Negatively charged hydrogen ions (H{sup -}), generated at the ion source, will be accelerated to 2.5 MeV. The facility is capable of producing one-ms-long pulses at 60 Hz repetition rate with up to 50 mA peak current. This beam will be safely terminated in the beam stop. In order to support the BTF project from a radiation protection point of view, neutronics simulations and activation analyses were performed to develop shielding around the beam stop, evaluate dose rates in case of an accident and to evaluate the radionuclide inventory at the facility end-of-life for decommissioning. The challenge of the analyses was to consider low energy charges particles reaction physics to fit the calculations into our routine transport and activation methods as practiced for the much higher energy spallation environment. (authors)

43 PARTICLE ACCELERATORS↗
NASA NTRS1994

Monte Carlo Radiation Analysis of a Spacecraft Radioisotope Power System

A Monte Carlo statistical computer analysis was used to create neutron and photon radiation predictions for the General Purpose Heat Source Radioisotope Thermoelectric Generator (GPHS RTG). The GPHS RTG is being used on several NASA planetary missions. Analytical results were validated using measured health physics data.

planetary missions spacecraft electrical power out↗
DOE OSTI2025

The scoping, design, and plasma physics optimization of the Eos neutron source stellarator

On the path to a fusion pilot plant, Thea Energy plans to build Eos, a sub-breakeven, deuterium-deuterium, beam-target fusion, stellarator neutron source facility for producing tritium and other valuable radioisotopes. In this paper, a set of 1D plasma physics models are coupled and used to design the operating point of the facility and predict performance. At this foundational stage of the design, analytic and approximate models are sufficient to capture the leading-order effects, and fast enough to run in the inner loop of an optimizer. Higher-fidelity analyses will follow. Models of 1D profile-dependent neutral beam stopping, ion beam slowing down, beam-target fusion, electron-ion classical heat transfer, energy confinement (ISS04), beam pressure, beam heating of ions and electrons, beam-beam fusion fraction, and neutral beam injection and gyrotron heating electrical efficiencies are included. A numerical optimizer is used to determine the minimum required facility electric power to generate tritium at a given rate. A potentially advantageous regime is described in which modern precisely-quasisymmetric stellarators, new high-temperature superconductors, ITER-derived neutral beam injection, and new high-frequency gyrotrons enable a suitible target plasma with hot electrons, cold ions, peaked density and temperature profiles, and high beam-injected ion density. It appears possible at this time for a facility with a medium-scale and medium-strength stellarator whose required facility electric power is less than 40 MW to produce $2.5\times 10^{17}$ neutrons s -1 for the production of radioisotopes. With the addition of a tritium breeding blanket, such a facility could produce 0.2 grams d -1 or 70 grams yr -1 of tritium.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS2017

Long-term Neutron Background Environment Measured by the Dynamic Albedo of Neutrons (DAN) Instrument onboard Mars Science Laboratory (MSL)

The Dynamic Albedo of Neutrons (DAN) instrument onboard Mars Science Laboratory (MSL) consists of a pulsed neutron generator (PNG) and the neutron detector/electronics (DE) module. There are two 3He-proportional counters used in the DE module. One counter is surrounded with a Cd shell enclosure and measures epi-thermal neutrons with >0.4 eV. The second detector does not have a Cd shell enclosure and measures both thermal and epi-thermal neutrons. The main scientific objectives of DAN are two-fold: (1) the primary objective is to measure the bulk hydrogen abundance (in forms of water or hydrated minerals) of the sub-surface and (2) the secondary objective is to measure the background neutron environment at the surface. DAN achieves the first objective by using a pulsed 14 MeV neutron source (ACTIVE mode). Emitted neutrons undergo a series of nuclear interactions with the regolith and lose energy. The DAN DE module then measures the time profile of the neutrons returned from the regolith. Even a small amount of hydrogen, as low as 0.1 weight % in the regolith, can effectively moderate the high energy neutrons [Mitrofanov, et al., ]. The less-moderated (or epi-thermal) neutrons arrive at the detector earlier than the more-moderated (or thermal) neutrons. The shape and magnitude of neutron time profile curves after pulses can be used to estimate the depth distribution of the hydrogen content. When no neutron pulse is used (PASSIVE mode) DAN measures the neutron background environment at the Mars surface. There are two sources of the background neutrons which DAN would measure: one is secondary neutrons generated by galactic cosmic ray (GCR) interactions with atmospheric and surface materials and the other is neutrons from the Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) onboard MSL as a power source. The passive mode data for the first 1300 sols of the MSL mission are presented in this paper. A summary of the DAN surface operation is described by [Mitrofanov et al., 2012; Jun et al., 2013.

Tate, C.↗
DOE OSTI2018

Dose rate verification for SNS spent structures - 25433

Residual dose rate measurements were conducted on two Spallation Neutron Source spent structures, target vessel no. 13 and proton beam window no. 5, after extraction from their service locations. These measurements were used to verify calculation methods for radionuclide inventory assessment that are typically performed for nuclear waste characterization and transportation. Neutronics analyses for predicting residual dose rates were carried out using complex and rigorous geometric models of the structures and their surroundings. Calculational results were analyzed and compared to the measured dose rates. (authors)

43 PARTICLE ACCELERATORS↗